• 전기의세계
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• v.22 no.1
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• pp.20-27
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• 1973

The multi-group diffusion theory is applied to the analysis of the currently constructing Kori reactor core which is to be refuelled by 3-region fuel loading pattern and also to the comparative study on a conceptually designed 5-region reactor core, under the condition that, apart from the thermal-hydraulic considerations, all the input data referred to here in are assumed to be identical for both cases. The numerical calculation is carried out for quantitative analysis of the characteristics of the two fuel loading patterns in details, and the calculated results show that, so far as the nuclear aspects are concerned, the characteristics of the 5-region reactor core are proved to be superior to those of Kori's 3-region reactor core in general.

• Nuclear Engineering and Technology
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• v.35 no.1
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• pp.45-54
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• 2003

A hyperbolic two-phase, two-fluid equation system developed in the previous work has been implemented in an existing nuclear safety analysis code, MARS. Although the implicit treatment of interfacial pressure force term introduced in momentum equation of the hyperbolic equation system is required to enhance the numerical stability, it is very difficult to implement in the code because it is not possible to maintain the existing numerical solution structure. As an alternative, two-step approach with stabilizer momentum equations has been selected. The results of a linear stability analysis by Von-Neumann method show the equivalent stability improvement with fully-implicit solution method. To illustrate the applicability, the new solution scheme has been implemented into the best-estimate thermal-hydraulic analysis code, MARS. This paper also includes the comparisons of the simulation results for the perturbation propagation and water faucet problems using both two-step method and the original solution scheme.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.21 no.12
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• pp.695-702
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• 2009

Dynamic behavior of Libr-water absorption system using low-temperature hot water was investigated numerically. Thermal-hydraulic model of single-effect/double-lift 100 RT chiller was developed by applying transient conservation equations of total mass, Libr mass, energy and momentum to each component. Transient variations of system properties and transport variables were analysed during start-up operation. Numerical analysis were performed to quantify the effects of bulk concentration and part-load operation on the system performance in terms of cooling capacity, coefficient of performance, and time constant of system. For an absorption chiller considered in the present study, optimum bulk concentration was found to exist, which resulted in the minimum time constant with stable cooling capacity. COP and time constant increased as the load decreased down to 40%, below which the time constant increased abruptly and COP decreased as the load decreased further.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.24 no.8
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• pp.1046-1055
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• 2000

The flow and thermal phenomena in flat tubes of radiator are analyzed numerically. To predict the characteristics of heat transfer and pressure drop, the flow analysis program for three-dimensional complex geometry is developed, which adopted an non-staggered grid system and Cartesian velocities as dependent variables of the momentum equations. Using the developed program, the effect of tube specifications on the heat transfer characteristics is investigated for various flat tubes. From this study, the following results are obtained; (1) For the same hydraulic diameter($D_h{\doteq}5.2$mm), the Nusselt numbers of three basic modeis(D, J, and H-model) are 8.71, 8.92, and 10.58, respectively, and the pressure drops of D-, J-, and H-model are predicted as $-3.08{\times}10^{-2}\;Pa,\;-3.12{\times}10^{-2}\;Pa,\;and\; -3.98{\times}10^{-2}$ Pa, (2) In case of the same flat tube specification, the fins must be brazed at upper tube surface because the heat is more vividly transferred. Therefore, it is found that the H- model is the most effective tube as a heat exchanger and these results are used as a fundamental data for the design of tube.

• International Journal of Automotive Technology
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• v.8 no.5
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• pp.659-666
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• 2007

In the present study, a numerical analysis of the three-dimensional heat transfer and fluid flow for a vehicle cooling system was developed. The flow field of the engine room between the grille and radiator was analyzed. The results show that, as the airflow inlet grille angle $\alpha$ is varied from $15^{\circ}$ to $-15^{\circ}$, the air flow rate compared with $\alpha=0^{\circ}$(horizontal) changes from -11.9% to +5.1%; while the heat flux from the radiator changes from -9.2% to +4.4%. When the airflow inlet bumper angle $\beta$ is varied from $-5^{\circ}$ to $+15^{\circ}$, the heat flux from the radiator compared with $\beta=0^{\circ}$(horizontal) increases up to +4.4%. When the airflow inlet grille angle $\alpha=-15^{\circ}$ and the bumper grill angle $\beta=+15^{\circ}$, the airflow rates and heat flux compared with($\alpha=0^{\circ}$, $\beta=0^{\circ}$) can be increased to +9.5% and +7.5%, respectively. The results indicate that the optimal angles for cooling efficiency are used.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2847-2858
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• 2021

When a nuclear reactor with rectangular fuel assemblies runs for a long time, impurities and debris may be taken into coolant channels, which may cause flow blockage, and the blocked fuel assemblies might be destroyed. Therefore, the purpose of this study is to perform a thermal-hydraulic analysis of a rectangular fuel assembly by STAR-CCM+, under the condition of one subchannel with 80% blockage ratio. A rectangular fuel assembly of the International Atomic Energy Agency (IAEA) 10 MW material test reactor (MTR) is chosen. In view of the gasket material taken into the coolant channel is close to the single side of the coolant channel, in the flow blockage accident of the Oak Ridge Research Reactor (ORRR), a new blockage category called single side blockage is attempted. The blockage positions include inlet, middle and outlet, and the blockage is set as a cuboid. It is found by simulations that the blockage redistributes the mass flow rate, and large vortices appear locally. The peak temperature of the cladding is maximum, when the blockage is located at the single side of the coolant channel inlet, and no boiling occurs in all blockage cases. Moreover, as the height of the blockage increases, the damage caused by the blockage increases slightly.

• Journal of the Korean Society of Mechanical Technology
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• v.20 no.6
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• pp.852-857
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• 2018

Vortex Generators are used in heat exchanger to enhance the heat transfer of air side. 3-D numerical analysis is performed on heat transfer characteristics of a channel with trapezoidal vortex generator. We investigate the effects of vortex generators with two different inclined angles to flow direction which are forward and backward vortex generators. The thermal hydraulic performance such as Nu and pressure drop, is compared quantitatively. The results show that vortex generator enhances the heat transfer by developing boundary layers and secondary flow in the downstream. The downwash flow region corresponds to the maximum Nu, while the upwash flow region corresponds to Nu minimum. In the view of the heat transfer characteristics, FVG is better than BVG. However, when flow is turbulent as Re increases, the pressure drop for FVG is higher than that for BVG.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.673-683
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• 2016

This paper analyzes the accuracy of the methods used in calibrating the thermal power of Nigeria Research Reactor-1 (NIRR-1), a low-power miniature neutron source reactor located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. The calibration was performed at three different power levels: low power (3.6 kW), half power (15 kW), and full power (30 kW). Two methods were used in the calibration, namely, slope and heat balance methods. The thermal power obtained by the heat balance method at low power, half power, and full power was $3.7{\pm}0.2kW$, $15.2{\pm}1.2kW$, and $30.7{\pm}2.5kW$, respectively. The thermal power obtained by the slope method at half power and full power was $15.8{\pm}0.7kW$ and $30.2{\pm}1.5kW$, respectively. It was observed that the slope method is more accurate with deviations of 4% and 5% for calibrations at half and full power, respectively, although the linear fit (slope method) on average temperature-rising rates during the thermal power calibration procedure at low power (3.6 kW) is not fitting. As such, the slope method of power calibration is not suitable at lower power for NIRR-1.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.29 no.12
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• pp.654-662
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• 2017

When emergency core cooling system (ECCS) is operated during loss of coolant accident (LOCA) in a pressurized water reactor (PWR), pressurized thermal shock (PTS) phenomenon can occur as cooling water is injected into a cold leg, mixed with hot primary coolant, and then entrained into a reactor vessel. Insufficient flow mixing may cause temperature stratification and steam condensation. In addition, flow vibration may cause thermal stresses in surrounding structures. This will reduce the life of the reactor vessel. Due to the importance of PTS phenomenon, in this study, calculation was performed for Test 1 among six types of OECD/NEA ROSA tests with ANSYS CFX R.17. Predicted results were then compared to measured data. Additionally, because temperature difference between the hot coolant at the inlet of the cold leg and the cold cooling water at the inlet of the ECCS injection line is 200 K or more, buoyancy force due to density difference might have significant effect on thermal-hydraulic characteristics of flow. Therefore, in this study, the necessity to include buoyancy force term in governing equations for accurate prediction of single phase thermal stratification in both cold legs and downcomer by ECCS injection was numerically studied.